This invention relates to a pressurized water cooled nuclear reactor and in particular to a flow path therethrough.
A pressurized water cooled nuclear reactor conventionally includes a core formed of vertically supported fuel element and vertically movable control rods passing therethrough. These controls rods are surrounded by guide tubes at least through the core to assure proper guidance of their movement. Flow of the coolant water is upward through the core to insure stability of flow in the event of any localized steam or overheating.
While the control rods contain no fuel they do absorb neutrons and thereby generate some heat. Cooling of the control rods is, therefore, required. The conventional method of cooling these rods involves passing a portion of the flow upwardly through the control rod guide tubes which may then exit either in the outlet plenum or in an upper portion of the reactor vessel from which location the flow passes to the outlet.
The flow passing through these guide tubes is in parallel with the flow actually passing over and cooling the fuel assemblies. It, therefore, must be severely restricted to avoid an undue reduction in the thermal performance of the core. This flow must pass through the guide tubes when the control rods are withdrawn as well as when they are inserted. The flow path has a relatively low pressure drop when they are withdrawn, and a concomitant increase in flow. In order to restrict these variations in flow, orifices must be placed at the inlet of the guide tubes. This cannot avoid the increase in core by-pass when the rods are withdrawn but it does minimize the extent to which the flow increases. The use of orifices involves not only the expense of installing these but also the potential of plugging which is inherent in any flow restriction which is put within a nuclear reactor.
The selection of the particular by-pass flow quantity through the control rod guide tubes requires a critical allocation of flow, since there must be sufficient flow to properly cool the rods in the fully inserted position, but any excess flow used needlessly degrades the thermal performance of the reactor core.
Since flow is upwardly along the control rods there is an upward force due to the drag of the fluid flow as well as the pressure difference between the bottom and upper portion of the control rod. The force resists the downward movement required in scramming a reactor, thereby lengthening scram time and increasing the forces required to drive the control rods down beyond what they would be in the absence of such flow arrangements.
In the conventional arrangement, the pressure below the core is higher than the pressure at the outlet of the core due to the friction drop of the flow passing therethrough. This results in a significant upward force on the core in the order of 3,000,000 newtons for a 280 kilopascal pressure drop. Since the entire upper portion of a conventional reactor vessel is at the outlet pressure this force can be resisted only by structures which transmit the force to the reactor vessel or reactor head.
In the conventional arrangement, the upper portion of the reactor vessel is not only at outlet pressure but also at outlet temperature. The core support barrel is the structure which separates the two pressure and temperature volumes. The support barrel is generally supported at the top of the reactor vessel body immediately adjacent the bolted joint between the head and the body. The complex structure in this area must not only tolerate the physical forces due to the internal pressure as transmitted through the bolts but must also simultaneously tolerate the thermal stresses due to the temperature difference on the two sides of the core barrel at the joint area.